Burnup Calculation of a CANDU6 Reactor Using the Serpent and MCNP6 Codes

M. S. Hussein; H. W. Bonin; B. J. Lewis

§ Publications 2014

Burnup Calculation of a CANDU6 Reactor Using the Serpent and MCNP6 Codes

M. S. Hussein, H. W. Bonin, B. J. Lewis

Type: Conference paper
Venue: 19th Pacific Basin Nuclear Conference
Place: Vancouver
Year: 2014

§1 Takeaway

What it did and showed.

Validated whole-core burnup in MCNP6 and Serpent against WIMS-AECL, the industry lattice code, with k-effective versus burnup in excellent agreement. Still cited in 2025 work on SMR spent fuel in CANDU.

His part: First author.

Part of: CANDU 6 core modelling →

§2 Read it

Where to find the paper.

IAEA INIS ResearchGate

Cited by: Mohsen et al., Nuclear Technology 2025 (doi 10.1080/00295450.2025.2553262).

§3 Cite

BibTeX.

@inproceedings{hussein2014burnup,
  author = {M. S. Hussein and H. W. Bonin and B. J. Lewis},
  title = {Burnup Calculation of a CANDU6 Reactor Using the Serpent and MCNP6 Codes},
  booktitle = {19th Pacific Basin Nuclear Conference},
  year = {2014},
  address = {Vancouver},
}

§· Related

On the same subject.

Related papers
No.	Year	Title	Venue
01	2014	Calculation of the Radial and Axial Flux and Power Distribution for a CANDU 6 Reactor with both the MCNP6 and Serpent CodesM. S. Hussein, H. W. Bonin, B. J. Lewis	PBNC 2014Conference · Vancouver